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The goal of this work was to develop a foil activation method to measure high-energy (1--120 MeV) neutron flux spectra at the Spallation Neutron Source by researching the scientific literature, assembling an experimental apparatus, performing experiments, analyzing the results, and refining the technique based on experience. The primary motivation for this work is to provide a benchmark for the neutron source term used in target station and shielding simulations Two sets of foil irradiations were performed, one at the ARCS beamline and one at the POWGEN beamline. The gamma radiation of the foil activation products was measured with a high purity germanium gamma-ray spectrometer, and the product reaction rates during irradiation were quantified. Corrections, such as self-shielding factors, were applied to the measurements to account for particular effects. The corrected measurement data, along with calculated response functions and an initial guess spectrum, were input to the MAXED neutron spectrum unfolding computer code. MAXED uses the maximum entropy method to unfold an output spectrum that is the minimally modied guess spectrum consistent with the measurement data. The foil irradiation and subsequent analysis from the ARCS spectrum produced a reasonable neutron spectrum, which noticeably differed from the initial guess spectrum. This measurement is regarded as consistent, but yet unverified. The gamma-ray spectrum of the foil irradiation at the POWGEN beamline showed no high-energy activation. This is regarded as an experimental error, and no conclusions can be drawn about the high-energy neutron spectrum. Future foil irradiations are planned to verify and expand the neutron spectrum measurements.
The qualitative aspects of the activation technique of neutron flux spectral determination are well known. The quantitative aspects are not so well known. In part, the failure of correlation and subsequent understanding of neutron radiation effect studies is attributed to a lack of adequate knowledge about the form of neutron environments. An Advanced Activation Method of Neutron Flux Spectral Determination based on the use of updated and more accurate cross section data and activation measurements is presented. This method will permit the definition of neutron flux spectral forms with increased accuracy for radiation effect studies. The results of error analysis studies on methods of foil activity data reduction and requirements on the accuracy of activation detector cross section data and activity measurements are presented and discussed. The appendixes of this volume contain tabulated data that are necessary for the performance of Step I of the advanced method of foil activation data reduction. Volume II of this report presents and discusses the updated evaluated cross section data. Volume III discusses the computer code required for the performance of Step II of the advanced method. Volume IV presents the results of Tory II-C flux spectral measurements using the advanced method of data reduction. (Author).
This book gives the state of the art in the field of reactor dosimetry as applied in nuclear power plants and research reactors. Surveillance programs are presented for nuclear power plants in Europe, including Russia and Ukraine, USA, Argentina and Korea. New cross-section measurements from most of the European, American and Japanese research reactors are reported. The latest developments in computer code development for radiation transport and shielding calculations, and radiation measurement techniques are also highlighted.
A survey was made of nuclear reactions which may be used in measuring neutron flux spectra and fluence. An important application of these data is the determination of neutron flux spectra and fluence for application to radiation effects studies (including dose to tissue). Published microscopic cross section data for 28 reactions have been compiled and are shown graphically. The evaluation of the microscopic data is shown by a line through the data. A tabulation of the evaluated cross section library is provided. Measured values of the resonance integral for selected reactions are reviewed and compared with resonance integrals calculated using the evaluated cross section library. The effects of 0.508, 1.016, and 2.032 mm thick cadmium are discussed. Average cross sections for boron-10 covered foils are calculated. The average cross sections of the 28 reactions in a fission spectrum is calculated. (Author).
A multiple foil activation iterative method has been developed to determine neutron flux spectra. A computer code (SAND II) has been written which provides a 'best fit' neutron differential flux spectrum for a given input set of infinitely dilute foil activities. The results of experimental and analytical studies for a wide variety of neutron environments strongly suggest that, in addition to determining neutron flux spectra, the multiple foil activation iterative method can be helpful in (1) the validation and improvement of calculational techniques used to predict flux spectra; (2) upgrading of confidence in neutron spectrometer measurements; (3) determining the reliability of foil activity measurements methods; (4) assessing material scattering/ absorption effects; and (5) examining current foil detector cross section evaluations to provide guidance for re-evaluation for these data, and eventual 'unfolding' of the absolute differential form of cross sections for any foil reactions producing a detectable product.